IR 05000213/1986099
| ML20237L673 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/02/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237L653 | List: |
| References | |
| 50-213-86-99, NUDOCS 8709090027 | |
| Download: ML20237L673 (66) | |
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s ENCLOSURE 4
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l U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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AMENDED
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SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE i
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INSPECTION REPORT 86-99
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CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK NUCLEAR POWER PLANT l
ASSESSMENT PERIOD: hARCH 1, 1986 - MARCH 31, 1987 BOARD MEETING DATE:
MAY 14, 1987
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SUMMARY OF RESULTS
'A.
Overall Facility Evaluation This SALP confirmed continued' safe operation although overall performance continued to decrease.
In 1982, all nine functional areas were rated as Category 1.
In 1986,. there were two Category 1 and eight Category 2 areas.
This SALP found one Category 1 area.(Security), ten Category 2. areas, and one Category 3 area.(Fire Protection). Actions to improve performance have been initiated by the new station superintendent (as-signed in' December 1986), but the effectiveness of these actions has not yet been demonstrated.
Notable improvements were evident in operator requalification, onsite-safety review committee performance, and control room utilization.
The licensee continued to exhibit a strong safety perspective and aggres-sively resolved those matters of immediate safety significance.
In ad-dition, risk reduction was actively pursued. When vulnerabilities were identified,' interim corrective actions were taken, and modifications were proposed and implemented.
Performance in the security area has remained excellent.
In this area, there was solid individual and supervisory performance, aggressive identification and resolution of problem areas, and a good interface with the corporate staff.
There were'seven plant trips during this 13-month SALP period.
Three of these were attributed to personnel error. None of the five trips during the preceding 12-month SALP period were due to personnel error.
These trips illustrate the need for improvement of both personnel and equipment performance.
Also, ALARA continues to be an area where im-provement is needed.
Radiation exposures remain exceptionally high and there was an overexposure that resulted in a civil penalty.
These mat-
.ters and procedure inadequacy and adherence problems marred a generally sound radiation protection program.
Site and corporate interface' deficiencies appear to be a major contribu-tor to performance weaknesses in a number of areas including fire pro-tection, engineering support, radiological controls, licensing, and out-age management.
In addition, other long-standing problems continue to affect performance. These include containment leak rate testing, the large open items backlog, numerous facility modifications, converting to standard technical specifications, and upgrading Safety Analysis documentation of design data.
In summary, while safe operation continues, there are numerous perform-ance weaknesses and an overall downward trend.
Recent initiatives are indicative of potential improvements, but tangible results to support a conclusion of improving performance were not available during this SALP period.
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B.
Background Licensee Activities On March 1, 1986, at the start of the SALP period, the licensee was com-pleting a refueling and maintenance outage.
The plant was restarted on May 6, 1986, after resolving an emergency core cooling deficiency which prevented long term core-cooling for a small break in the charging system piping. Also, the licensee identified an unplugged, degraded steam generator U-tube and was granted temporary, one-cycle relief for opera-tion with the tube not plugged.
Full power was achieved on May 22.
During the SALP period, seven plant trips occurred because of personnel
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errors, equipment, and other maintenance related failures.
Numerous turbine load runback actuations resulted from nuclear instrumentation system (NIS) fluctuations caused by aging components within the NIS.
The plant shut down for 20 days in July 1986 to plug steam generator U-tubes in which defects were identified in the tubesheet area.
The ap-parent cause of the defects was the tube rolling process during fabrica-tion and subsequent primary side stress corrosion cracking.
During this outage, a steam generator worker exceeded his quarterly radiation expo-sure limit.
In December 1986, the licensee identified another emergency core cooling system design discrepancy affecting long term core cooling in the sump recirculation mode.
Interim corrective actions involved a plant shutdown on December 19 to perform a special flow test to verify the acceptability of the proposed throttling of injection flow.
The plant returned to power on December 25, 1986.
Overall, the plant achieved a capacity fac-tor of 63% during this SALP period.
The plant's lifetime capacity factor is 76%.
A more detailed description of plant activities is provided in Section V.F of this report.
Inspection Activities Two NRC resident inspectors were assigned to the site during the assess-ment period.
The NRC inspections are summarized in Table 1 and represent an inspection effort of 3590 hours0.0416 days <br />0.997 hours <br />0.00594 weeks <br />0.00137 months <br /> (3314 hours0.0384 days <br />0.921 hours <br />0.00548 weeks <br />0.00126 months <br /> per year), distributed as shown in Table 2.
Special team inspections were made of licensee ALARA practices (April 7-11, 1986); the annual site emergency exercise (April 25, 1986); licen-l see implementation of Appendix R fire protection requirements (June 16-20,
1986); and overall facility operation (November 14-21,1986).
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Twenty-one enforcement action.s, including a Severity Level III violation
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and Civil Penalty for the radiation overexposure, were issued.
Enforce-ment actions are tabulated in Table 3.
C.
F_acility Performance Tabulation CATEGORY CATEGORY LAST THIS FUNCTIONAL AREA PERIOD *
PERIOD **
TREND 1.
Plant Operations
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Maintenance & Modifications
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Surveillance
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Fire Protection
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Engineering Support
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Licensing Activities
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Refueling / Outage Management
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Radiological Controls
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Security & Safeguards
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11. Training and Qualification
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Effectiveness 12. Assurance of Quality
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March 1,1985 to February 28, 1986
March 1, 1986 to March 31, 1987
Not addressed as a separate area.
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Unplanned Trips and Shutdowns j
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ROOT FUNCTIONAL-
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.DATE LEVEL
' DESCRIPTION CAUSE
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AREA
'5/8/86 10%
Automatic trip due to load-Personnel error -
Operations increase past 10% power (P7) operator exceeded during main turbine testing P7 with an MSIV with the #2 main steam iso-closed.
lation valve (MSIV) closed.
6/4/86 100%-
Anticipatory manual trip due Random equipment
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to prompt (and proper) opera-failure - LCV tor response to feedwater operator-to-stem system fluctuations.
Failure failure.
of a heater drain tank level'
control valve (LCV) caused rapidly decreasing steam
_ generator levels.
6/17/86 100%
Anticipatory manual trip Inadequate failure Maintenance-similar to the trip on analysis.
6/4/86. The same LCV oper-ator separated again. The cause of the previous LCV failure was not adequately identified therefore approp-riate actions to prevent re--
currence were not implemented.
6/19/86 0%
Automatic trip while trouble-Personnel error-Maintenance shooting a blown fuse in a inadequate control nuclear instrument power of maintenance.
supply. The fuse was re-placed without determining the initial cause of failure or assessing the potential effects of reenergizing a faulted. component. The shorted
. power supply overloaded the vital instrument bus and un-blocked the existing closed MSIV reactor trip signal.
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POWER ROOT FUNCTIONAL DATE~
LEVEL DESCRIPTION CAUSE AREA 6/22/86 10%'
' Automatic trip during 3-loop Inadequate control Maintenance operation. Ceiling debris of cleanliness
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lodged in a reactor protec-during maintenance.
l tion system (RPS) relay after
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replaced. This created an un-(
annunciated low flow half-trip i
signal.. A second half-signal
was_ actuated by the 3-loop.
operating condition. The plant tripped when the low flow trip signal was automatically un-blocked as turbine load ex-ceeded 10% power.
7/11/86 100%
Shutdown to mode 4 to re-Recurrent plant Maintenance j
place a sheared charging shaft failure due
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pump shaft. The licensee was to untimely failure unable to affect repairs analysis and correc-within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. grace tive action in re-period and placed the unit sponse to an earlier in hot shutdown.
shaft failure.
The plant remained shutdown until August 5,.1986 for steam generator repairs described below.
7/15/86 0%
Three week outage to plug 119 Construction
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steam generator tubes.identi-defects and tified as having flaws in the primary side stress rolled region of the tubes.
corrosion cracking.
8/30/86 100%
Anticipatory manual trip due Random equipment
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to prompt (and proper) opera-failure - defects tor response to a failed open in the FRV stem-to-feedwater regulating valve plug weld.
(FRV).
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POWER ROOT FUNCTIONAL DATE LEVEL DESCRIPTION CAUSE AREA 11/30/86 100%
Automatic trip due to low Personnel error-Maintenance steam generator level caused inadequate control by inadvertent closure of of maintenance.
the #3 FRV. Technicians and operations supervision did not assure that the FRV control system was deenergiz-
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ed prior to work. FRV control power supply was shorted out causing the valve to fail closed.
The plant remained shutdown for 4 days to repack a reactor coolant loop isolation valve which leaked following this trip.
12/6/86 25%
Shutdown to remove vibrating Design error.
Engineering moisture separator reheater Support baffle plates which were in-stalled during turbine over-hauls. Thermal expansion stresses broke away the baffle plate mounts.
12/19/85 16%
Shutdown to perform a special Design / analysis Engineering flow test to verify the ade-error.
Support quacy of emergency core cool-ing systems (ECCS) due to a design deficiency affecting the recirculation mode after a medium sized LOCA.
NOTE: The root causes in this table are the SALP Board assessments based on in-
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IV.
PERFORMANCE ANALYSIS A.
Plant Operations (1270 hours0.0147 days <br />0.353 hours <br />0.0021 weeks <br />4.83235e-4 months <br />, 35%)
1.
Analysis The last SALP rated this area as Category 1, consistent.
No major concerns were identified.
Events involving personnel errors, the quality of submittals to the onsite safety review committee (PORC),
and the effectiveness of root cause addressal were highlighted for licensee evaluation.
Corrective action effectiveness was shown during the current SALP period by a reduction of items such as fire barrier problems and late procedure reviews. As discussed in the following text, personnel errors and quality of submittals to PORC are still weaknesses.
This assessment is based on routine resident inspections throughout the SALP period and a Region I Diagnostic Team Inspection conducted in November 1986.
A capacity factor of 63% (76% lifetime average) resulted from an extended refueling outage, two mid-cycle shutdowns to address steam generator and emergency core cooling system deficiencies, and seven plant trips. One of these trips occurred due to operator error during plant startup.
The operators did not maintain power below 10% to assure that a bypassed main steam isolation trip signal was not automatically reinstated. Two other trips occurred during troubleshooting of instrumentation problems. Better coordination between the operators and technicians could have prevented these two trips.
Operating shift activities were conducted professionally.
Shift turnovers were thorough.
Log keeping was adequate.
Activities were generally well understood, and were conducted with care and formality.
Operators responded effectively to equipment failures and plant
trips. Operator alertness was evident during dayshif t and backshift i
inspections.
Distractions such as extraneous reading material were not permitted in the control room.
The licensee rearranged the control room to remove unnecessary administrative materials and provide more space away from the operator-controlled area for opera-tor aides and for equipment tagging activities. Also, the licensee initiated modifications to expand the control room to provide seg-
regated space for the shift supervisor.
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The licensee has effectively minimized operator distractions due
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to illuminated main control board alarms.
During routine operations there are normally only one or two lighted alarms, and projects have been initiated to correct these indications.
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L 0perator technical knowledge was good.
NRC license examination re-
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suits (14 of 15 passed) showed that the candidates were well trained for initial = license examination. Operator interviews by inspectors and NRC participation in the operator. requalification program demon-strated that the operators were knowledgeable.
The NRC has been concerned about weaknesses in the operator requali-
.fication program since 1984.
Progress toward a new continuous re-
. qualification program had been slow.
During this SALP period, NRC reviews of the requalification program found satisfactory training-effectiveness.
However, some program administration details had
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not been' formalized.
For instance, several draft lesson plans.were in use, and there was no method for documentation of removal of operators from shift due to poor performance in requalification.
Also, the licensee's formal requalification program submittal did not specify the attributes of an annual comprehensive written ex-amination, and no formal program evaluation / audit was specified.
These factors indicate that more management attention to program details is appropriate.
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Management involvement in operations was evident in routine super-intendent tours of the contro'1 room and plant spaces, in regular plant material inspections by superintendents and department super-visors, and in active and assertive participation in daily manage-ment meetings. However, management was not effective in resolving a problem involving several instances of manual containment isola-tion valves being opened in violation of Technical Specification (TS) requirements. While the safety significance of these viola-tions was small, the failure to recognize the need for strict com-pliance with TS requirements and propose prompt and comprehensive corrective actions resulted in repeat violations.
The Plant Operations Review Committee (PORC) and the Nuclear Review Board functioned well.
Inspector observations characterized these meetings as frank technical discussions with full management support for the expression of dissenting views. PORC met more than once a week to complete the large volume of procedure, operational event, modification, and Technical Specification reviews.
The licensee made some progress in reducing the PORC workload by screening out minor event reviews which are adequately handled by line management, and by use of subcommittee pre-reviews for large packages.
Notwith-standing, the quality of staff work prior to PORC review'has re-mained inconsistent. While the general quality of modification packages has improved, several packages were returned for documen-tation of the bases for acceptability. Also, the continuing large number of procedure changes submitted to PORC indicated that the previous review process has not been properly effective. The good Nuclear Review Board (NRB) involvement was particularly evident in the two safety analysis / emergency core cooling system design errors identified during this period. Detailed NRB review and their in-
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quisitive approach to these problems contributed effectively to satisfactory interim resolution.
The NRB also noted inadequate correction of the root cause in selected Licensee Event Reports and initiated corrective actions.
NRC review of Licensee Event Report (LER) quality based on ten selected LERs resulted in the-same overall acceptable rating as in the previous period.
Improvement.was noted in safety consequence discussions and failure mode determinations. Weaknesses were noted in identifying the dates and times of major occurrences, in failed component information, and in cause and corrective action summaries in-the abstract section. During this SALP period, LERs improved and were more consistent due to the licensee's formalization and track-
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ing of LER development and' commitment implementation. This action resulted from NRC and licensee identified weaknesses in previous LERs, and from: licensee identification that an LER commitment to prohibit'3-loop operation in the startup mode had not been imple-mented. This resulted in plant operation outside this temporary limitation. No similar problems occurred under the new tracking system. The high proportion of personnel error-related events re-mained about the same during this SALP period as it was before.
While LER improvement initiatives were evident, these error-related events and other instances where LER corrective action was not ade-quate indicate that further improvement in root cause analysis and corrective action is warranted.
Plant housekeeping generally improved during this period.
The lic-ensee continued efforts to expand facilities for radioactive waste storage and employee habitability.
Early ia the period, there was an overall appearance of cleanliness but a lot of clutter due to the uncontrolled storage of materials near equipment and atop cabi-nets. An improvement was made in this aspect, but a formal program for control of equipment storage was not implemented.
Improved cleanliness was also noted in the auxiliary feed pump area.
In con-trast, efforts to recover areas which had become dirty or contamin-ated during the refueling outage did not extend into historically unkempt areas such as the RHR pit and pipe trenches. Overall, housekeeping was satisfactory.
In summary, improvements were noted in operator requalification, PORC performance, and control room utilization.
Steps were taken to address correction of recurrent weaknesses.
Further improvement in the quality of input to the PORC and the effectiveness of root cause analysis and corrective action for error-related events is warranted.
For the most part, plant down time was to correct iden-tified equipment problems and reflected proper response to the associated conditions.
However, operator error caused one trip and operator performance contributed to two other trips.
Specific licensee action to reduce the frequency of unnecessary trips is needed. Overall, this SALP found satisfactory operating performance.
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Conclusion Category 2.
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Board Recommendation
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Licensee: Correct the causes of operator contribution to unnecessary I
trips.
NRC:
None.
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Maintenance and Modifications (425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br />, 12%)
1.
Analysis
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The.previousESALP rated Maintenance as Category 2, consistent. Con-
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cerns included problems in the implementation of plant modifications, poor documentation of maintenance, and the maintenance backlog.
-Improvements were noted in these areas.during this period, but those.
problems are still a concern.
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The licensee.has a strong preventive maintenance' program. Activi-ties are tracked-and scheduled through the computerized Production Maintenance Management System (PMMS), and preventive maintenance
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items on both safety-related and balance of plant components are routinely completed.
In one case, the licensee was not sufficiently sensitive to the time consumed by maintenance.
Inadequate pre-
. planning. prior to removal from service contributed to the diesel
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fire pump being out of service on three consecutive days for work that is normally completed in one day.
Overall, however, the ef-festiveness of preventive maintenance was reflected in good equip-ment reliability.
No plant outages or safety-related equipment inoperabilities were attributed to preventive maintenance weaknesses
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during this SALP period.
Corrective maintenance was generally well-controlled.
Identified equipment discrepancies were entered into the PMMS system and auto-matica11y converted to work control documents.
In October 1986, the licensee initiated a new program to develop automated PMMS tracking of equipment spare parts listings, bills of materials, and recommended parts stock levels.
Daily planning meetings at manage-ment and supervisor levels contributed positively to the coordina-tion of repair activities. The licensee reduced the backlog of corrective maintenance actions by about 17-percent, but over 300 outstanding work items remain open. Many of these await spare parts,
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engineering review, or a plant shutdown.
The maintenance staff has effectively managed action items, assuring that safety significant work is completed.
During outages, an inter plant maintenance force (IMF) is available to augment the licensee's staff.
However, staffing levels appeared strained by increased commitments to support new technical training programs implemented late in this SALP period and to provide maintenance assistance at other plants. The licensee should take care to assure the reduction in backlog continues despite training and IMF commit-ments.
NRC observations of maintenance and discussions with workers and p
supervisors identified competent and knowledgeable personnel. Job
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supervisors were routinely present in the field and were observed to provide valuable assessment of technical problems.
However, job
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- oversight was noted to be lacking in radiation exposure control (see Functional Area H of this report) and there was inadequate documentation of maintenance.
For example, a violation was cited
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replacement of electrical: leads for an environmentally qualified motor-operated valve.
Licensee quality assurance audits have iden-tified similar problems in attention to detail and administrative procedure compliance.
Licensee management has emphasized procedural
- compliance and careful documentation in departmental meetings and written guidance to the staff.
So far, observations;have not found the results to be effective overall.
The. licensee has implemented expanded technical training programs for maintenance personnel. These programs also include maintenance team training for maintenance, health physics, quality control and supervisory personnel. The teams repaired full-sized mock-ups such as reactor coolant pump seals to improve the repair procedures, worker effectiveness, and job exposure controls.
Inspector discus--
sion of these training activities with workers revealed highly positive evaluations of the quality of this program.
The programs are relatively new, however, and were not effective in preventing the maintenance-related errors discussed below.
In addition, in-creased training commitments contributed to the backlog of out-standing work i' as.
Two plant trips durl this SALP period occurred during instrumen-q tation troubleshootii.g.
In one case, the technician failed to de-
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energize the circuit on which he was soldering.. The plant tripped when the soldering iron shorted the circuit and closed a main feed-water regulating valve.
In another case, a failed nuclear instru-ment drawer was re-energized without assessing the potential con-sequences,.and the resultant surge on the vital bus unblocked an existing plant trip signal.
Four other plant trips involved equip-
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- ment problems or failure to correct the root cause of a previous
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component failure.
Two of these were assessed as due to random
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equipment failure.
Another occurred when the heater drain tank
level control' valve failed closed, causing a loss of feedwater and
low steam generator' levels.
The valve operator-to-stem coupling had disengaged.
This problem had occurred several days before (one
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of the random failures), but repair was not effective in preventing the second trip.
After further repair (addition of set screws to hold the coupling in place) the problem did not recur. Also, a trip occurred because debris from the control room ceiling, which was replaced during the 1986 outage, became lodged in a reactor protec-tion system low flow trip relay. Maintenance controls during the q
ceiling replacement and subsequent Halon discharge tests did not
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. protect the reactor protection system from falling debris.
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debris held a relay contact open and created an unannunciated half-trip. When the plant was placed in 3-loop operation, a second half-trip was generated and the plant tripped.
These events point out a need for better maintenance control.
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Fifteen equipment failure related events were reported during this SALP' period.
Eight of these involved design and aging problems in the low temperature overpressure protection and nuclear instrumen-
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L tation. systems, respectively.
The licensee has projects ongoing
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'to~ correct these problems by.1989.
A recurrent charging pump shaft failure also occurred during this period.
NRC review.of this repair a'nd previous failure data con-cluded that the failure mechanism was'not apparent during the first
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repair and that vendor _ assistance did not promptly identify effec-L Ltive corrective actions. Vendor recommended corrective actions to lower. axial clearances were subsequently incorporated into_ charging pump maintenance procedures, but ao' action was taken-to verify con-
.formance of-the' initial repair to these new' pump clearances. After the second failure, measures recommended by the vendor to reduce cyclic stresses were implemented. These measures have been effec-
-tive to date.
.In summary, the maintenance and maintenance training programs are good.
However, parsonnel errors, other instances of ineffective maintenance, procedural.and documentation inadequacies, and the maintenance backlog were assessed to be factors which prevented achievement of overall maintenance excellence ~.
2.
Conclusion Category 2 3.
Board Recommendations Licensee: Consider measures to improve equipment and maintenance personnel performance in order to prevent unnecessary trips.
NRC:
None.
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.17 Cr Surveillance (460 hours0.00532 days <br />0.128 hours <br />7.60582e-4 weeks <br />1.7503e-4 months <br />, 13%)
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An'alysi s Surveillance was rated Category 2 during the previous SALP.. Con-cerns included surveillance procedure adequacy, missed surveillance, QA adequacy, and correction of long standing containment leak rate.
testing issues.
During the current period, licensee progress in these areas was noted.
However, problems. identified indicated a need for further improvement.
This assessment is based on six region-based NRC inspections and frequent resident inspector observations of surveillance.
An effective and generally well controlled surveillance program is in place using computerized scheduling and tracking programs. The
. program is. divided. into surveillance (SURs) for Technical Specifi-cation (TS). required testing and preventive maintenance (PMPs) for other equipment.
Both the SUR tracking system and Production Main-tenance Managenent System (PMMS) effectively scheduled surveillance requirements on a weekly basis.
No missed surveillance were iden-tified. There was, however, one NRC-identified instance where new TS requirements for radiological chemistry surveillance were not-properly incorporated in the SUR tracking system. Also, these tests were not completed. The licensee is evaluating changes necessary to improve commitment documentation and tracking.
The licensee has completed a 2 year effort to upgrade surveillance.
During this-SALP period, the primary thrust of this program has been the upgrading of procedures.
NRC inspectors noted improved quality in many surveillance procedures, particularly those for inservice testing of pumps and valves. Acceptance criteria were included in those procedures and operators performing the tests were thereby alerted to test failures.
The NRC also noted inconsistencies in procedure format within departments. These indicate that the sur-veillance upgrade program was not fully successful.
The licensee recently initiated a program to standardize the format and improve procedure details for technicians in the field.
No plant trips were caused by surveillance testing errors. There were two minor NRC violations for using an alternate data measure-ment technique and an uncalibrated test gauge.
The licensee promptly corrected these problems.
Other surveillance program deficiencies including delayed test result reviews, informal documentation, in-adequate independent verification of a jumper / bypass, and an un-approved procedure change to correct test connection locations for a valve interlock test.
The number and variety of these problems indicate that formality of implementation and attention to procedure details, particularly at supervisory levels, needs improvement.
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During the 1986 refueling outage, the licensee performed eddy cur-rent testing of almost all of the steam generator U-tubes.
The eddy current test program was improved through the use of better equipment and technology.
This was evident in the licensee's 1 den-tification of U-tube flaws in the rolled area of the tubes.
These defects were related to construction process deficiencies and pri-mary side stress corrosion cracking at the stressed locations.
Upon identification of the problem after the outage, the licensee shut down the plant, proposed acceptable defect acceptance criteria for NRC approval, and plugged 119 tubes which exceeded the criteria.
There was one problem related to eddy current testing (ECT) during this period.
Poor communication and verification of ECT data re-sulted in one tube with a 55% through wall defect not being plugged during the refueling outage.
The licensee identified this problem during a subsequent review of the steam generator inspection report.
After justification was developed, the NRC approved the licensee's request for leaving that tube unplugged during the current operating cycle.
The post outage startup physics testing program included precritical tests, zero power physics tests, and power ascension tests.
NRC inspectors observed effective management involvement.
Reactor en-gineering staffing was ample to assure safety.
Reactor engineers were found knowledgeable in their assigned areas.
New trending /
testing computer programs were developed.
For instance, monitoring rod drop measurements by computer permitted acquisition of data from all rods in a bank at one time versus a single rod drop using the old method.
The reactor engineers were found to interface effec-tively to accomplish their tasks.
For example, they provided the operating staff with proper updated graphs and figures and obtained input from the chemistry group for fuel integrity evaluations.
Containment leak rate testing (CLRT) was assessed based on NRC ob-servation of both local and integrated leak rate tests and problems in resolving outstanding CLRT licensing issues. An integrated CLRT was performed during this SALP period. The test procedure and methodology were much improved since the last test, yet several procedure weaknesses such as nonspecification of the start time and stabilization period for the verification test were noted. Manage-ment involvement was considered to be deficient, in that the test director was generally uninformed about the progress of integrated CLRT preparation and was not knowledgeable of the location of the verification test flow meter.
Licensee personnel conducting the test were found to be knowledgeable and competent.
Staffing levels l
were adequate during the CLRT.
As found containment leakage exceeded allowed limits for the second consecutive integrated test and the third consecutive set of local I
tests.
Both failures were due to excessive penetration leakage.
The licensee plans to modify several penetrations during the 1987
-_
____ _-__.
_ _ _ _ _ _ - - _
nm Li
,
y-K.;_
'
'
refueling outage,' including the most recurrent leaking penetrations.
NRC noted improved performance in.the conduct and frequency of.'l.ocal CLRTs.
Procedures were more detailed, particularly regarding test acceptance criteria and: prompt notification of test results. This had been a' weakness in previous CLRTs.
The_ licensee'also modified the component cooling water system by' adding a slipstream filter
.to reduce the impurities which had 'previously clogged the contain-ment isolation valves in this system. Subsequent tests indicated-
-
improved leak. rate performance, but the success of this change will.
-
' depend on the as-found leakage following the current operating cycle.
Several outstanding CLRT licensing issues remain open. Overall, CLRT failures indicate that additional licensee attention to main-tenance, modifications, and/or more frequent testing is needed to assure. acceptable containment leakage throughout operating periods.
Quality assurance (QA) involvement in the surveillance program was evident during'this period.
Expanded QA observations of field ac-tivities included coverage of CLRT and startup physics testing.
Licensee findings related to procedure inadequacy / noncompliance 7upported the NRC conclusion that further management attention to
.these areas.is necessary. The NRC'also noted that quality assurance
. group reviews of TS surveillance documents failed to identify tardy supervisory reviews and procedural deviations.
That indicates weakness in the quality of these QA reviews.
In summary, the surveillance program was technically sound and pro-perly~ scheduled. Weaknesses regarding procedure adequacy,. minor noncompliance, and attention to detail in documentation have been tdentified.by-the licensee and NRC. These indicate that previous upgrade programs were not fully effective. As-found' leak rate testing results indicate that acceptable containment leakage has not been-assured throughout operating cycles.
Corrective action programs related to these problems are ongoing, but their effec-tiveness has not yet been evident.
Quality assurance involvement was improving, although.the quality of document reviews was weak.
2.
Conclusion Category 2 3.
Board Recommendations Licensee: Assure that acceptable containment leakage is maintained throughout operating cycles.
NRC:
None l
_._i__.__._____.._______..__-
-_
_ _ - _ _
+
'20
.._
D.-
Fire Protection (276_ hours, 8%)
1.
Analysis Fire protection was part of the Plant Operations area in the pre-vious SALP. The frequency of reported fire protection system de-ficiencies was highlighted for licensee attention.
This concern was addressed by.the licensee through personnel indoctrination and
' training, _ and was not a concern during the. current SALP-period.
This area. includes fire prevention, fire detection, the ability to respond to fires with onsite and offsite forces, engineering and hardware features to limit fire spread, and.the ability to achieve safe shutdownLin the event of fire. An NRC team inspected the im-pigmentation.of NRC fire protection requirements (10 CFR 50, Appen-dix R) during this SALP.
Licensee performance in this area was routinely observed by the' resident inspectors.
The onsite fire brigade was well equipped and trained, with regular drills and classroom training for each brigade member.
During this SALP period, the fire brigade responded effectively to several small fires within the owner-controlled area.
Fire fighting equipment was generally maintained in good working order through inclusion in plant surveillance and preventive maintenance programs.
In one case, however, the NRC identified omission of monthly checks on a few portable fire extinguishers.
Also, the licensee identified an
.
instance where a fire suppression system was rendered inoperable when the actuation system failed due to exposure to the outside-environment.
The actuation circutt was rebuilt but the environ-mental cause was not addressed. Offsite committee review of the Licensee Event Report on this failure initiated-further action to-protect the system from the outside environment.
NRC review identified weaknesses in site awareness of fire protec-tion program commitments and in program management of the Appendix R upgrade effort.
For' instance, the safe shutdown analysis gene-rated by-the licensee lacked thoroughness and accuracy.
Systems necessary for safe shutdown (such as component cooling) were not included in the_ Fire Protection Evaluation Report. Critical calcu-
'lations for a cooldown analysis contained errors because the calcu-lations failed to take into consideration actual as-built conditions.
One deviation and four violations were identified.
Plant staff and engineering personnel in the company's headquarters did not appear to communicate effectively. As a result, commitments made to the NRC regardi.ng electrical breaker control were not fully implemented and procedures lacked critical elements such as accept-
!
ance criteria and an applicable setpoint change.
In addition, the j-licensee's interim procedures for shutting down following a control L
room fire and evacuation did not provide for monitoring some im-l.
-
-
.
-
- (
- =
portant process' parameters such'as-steam generator pressure and j
reactor, coolant loop temperatures. Also, the'NRC identified weak-
)
nesses-in emergency lighting.
The licensee subsequently. upgraded
]
the instrumentation and. lighting for'.these evolutions.
A'. lack.of thoroughness also appeared in some modification work.
The'results of the control room Halon suppression system discharge test were not adequately analyzed and'dispositioned. The NRC found-
'tha.t the test revealed that under certain plant conditions the Halon
!
' system would not' provide the appropriate level _of fire protection.
l
,
After NRC identification, the licensee took action to prevent those conditions which degraded Halon system performance. Another. problem was that plantimodifications were made.without taking into consi-deration the effect of these changes on'other safety-related equip-ment.
A. gas; bottle for.' remote operation of containment isolation valves was lashed to a conduit in-the auxiliary building without'
!
evaluating the consequences-of this action.
Also, a wheeled breath-
'
ing-air cart (approximately 1500 lbs) was located in the control room;without' installing appropriate seismic restraints. The licen-see began implementation of a' comprehensive equipment control and storage program.
Discrepancies were identified in the previous installation of fire barriers in the control room and switchgear room.
In response, the licensee committed to review commitments from previous NRC fire-protection evaluations and verify continued maintenance and control of these items.
This project is ongoing.
Although~the licensee was aware of the hands-on training expected for fire watches (these requirements were identified at an inspec-tion of another of the licensee's facilities), the licensee did not provide this training at Haddam Neck until it became an issue during an NRC inspection.
With regard to general fire protection concerns, the licensee had no designated person dedicated to fire protection managemerit at the plant. As a result, some fire protection issues did not get the attention that would be expected if a dedicated person were present onsite.
For instance, local storage of lube oil in unauthorized containers throughout the plant was not corrected until NRC identi-fied this problem.
During this SALP period, however, the licensee initiated weekly site inspections by a corporate fire protection engineer in company with site supervisors.
These inspections. pro-vided needed fire protection training onsite, as well as timely identification of deficiencies in this area.
In addition, a per-manent fire protection engineer position was authorized in January 1986 and is to be filled prior to the July 1987 refueling outage.
l
.
..
_ _ _ _
_ __ -
e
,
I In summary, fire brigades were well trained and equipped to fight on site fires.
Significant fire protection program weaknesses were identified in safe shutdown analyses, in equipment for shutdown outside the control room, in emergency lighting, in fire watch training, and in evaluation of plant modifications. These problems reflect a lack of management involvement in this area and inadequate coordination of fire protection activities between the site and corporate staffs. However, fire protection inadequacies did not cause inadequate fire safety.
2.
Conclusion Category 3.
3.
Board Recommendation Licensee: Take appropriate action to improve management involvement in the fire protection area and enhance the interface between the site and corporate staffs.
NRC:
None l
l
_.. _ - - _ _. '
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-
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4 i i
.
.,
E.
' Engineering; Support
!
'W 1.
LAnalyshr.
.
U
-
,
Thisisanewfunctionalarba.
It,is based on observations made.
during resident and specialist inspection of technical activities outside those provided by th'e operations, maintenance and instru-g mentation and controls (I&C) departments.
T h
The licensee maintains a site engineering group of.about 20 engi-neers and' technicians. Additional engineering support is available A@
from a large corporate support organization (NUSCO) which services-9*
'
4 nuclear units and several fossil sunits.
The. site engineers are relatively inexperienced and have: had a high turnover (9) during this period. They are responsible for site. implementation of. design changes, and for. engineering evaluation'of operational' problems.
",
More detailed engineering, design and analysis work is requested from the NUSCO organization.
Site engineering also includes reactor engineering personnel ;to monitor reactor performance during opera-tion, to coordinate refueling activities, and to conduct startup physics tests.
Inse'rvice inspection and testing personnel are also assigned to the site engineering staff.
"
In most technical disciplines, the large NUSCO support organization has the engineering knowledge and experthe needed to support the i
site. NUSCO supplies design and project management _ functions for-plant modifications, performs engineering.e' valuations' of site prob-lems as requested, and manages reactor fuel design and, plant safety analyses.
NUSCO also has. specialty groupswhich conduct containment
' leak rate testing, monitor plant equipment reliability, develop.
environmental quaMfichhn (EQ) programs, and provide probabilistic risk assessment (PRA).
NUSCO PRA support has contributed positively to safety during the
>
current SALP period.
Loss of offsite power was found to have a l
potential for resulting in loss of the emergency diesels due to
,
cooling water valves failing shut. Changes were implemented to.
<
assure that theSvalves are open to support diesel operation. Also, l
as a result of' PRA review, additional electrical separation measures
'
s.
are being inc6rporated in the design of the new switchgear room i
which '
- cheduled for completion in 1989.
In addition, P.RA was i r, in identifying two emergency core cooling system design p robl en..
. -ing certain postulated loss of coolant accidents (LOCAs) in the charging system or core deluge piping, long term core cooling in the recirculation mode was not assured. Upon identifi-cation of these problems, the engineering support organization an-alyzed tne safe 4 consequences and promptly proposed interim actions which were approved by NRC and implemented at the plant. One aspect of these changes involved throttling a valve in the residual heat removal system to prevent loss of safety injection flow through a
_ _ - - - _ _ _ _ _ -
_
.
.
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wh, M.)
'
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( C(
core'd$logesystembreak.
The licensee's proposal to throttle the val've without a confirmatory flow test was not accepted because of
~
the sensitivity.of flow-rate to valve position. The NRC required
.gVa flow test to be conducted after. setting the valve position.
Not-anthstanding this testing weakness, PRA-related engineering support t
"his bein:an excellent contributor to assuring and improving facility M
safety.
,
The licensee is committed to maintaining the Haddam Neck PRA up-to-date in regard to new plant modifications.
PRA review of a modifi-
',
cation no provide a nitrogen blanket on the demineralized water storage tank (DWST) identified that the modification increased risk because the probability of DWST collapse due to breather valve fapure was high. The licensee disabled the nitrogen blanket system Aefore'it was to be placed in service. While the identification 4f this problem shows good performance and diversity within the-licensee's organization, it also indicates a potential weakness in engineering this modification without sufficiently addressing the reliability of the breather valve design.
NUE0 developed the environmental qualification program for elec-T tr_ical equipment. During the 1986 refueling outage, many component deviations from the test report configurations were identified.
These discrepancies apparently resulted from inadequate field veri-fication of various component configurations.
NUSCO provided ef-fective engineering services to identify and correct these problems prior to plant startup. This occurrence indicated a lack of co-
' ordination between.the site and corporate organizations.
Similar j
coordination problems were noted in the implementation of fire pro-
,",
tection program upgrades (see the Fire Protection Functional Area).
y Engineering support for pla'nt modifications was inconsistent during this period.
The technical design work and safety reviews were generally good. Many modification packages from the 1986 refueling t
g'
outage were completed satisfactorily.
In contrast, inadequacies in post-modification testing for the Apsendix R shutdown indication panel and a containment isolation valve replacement, and material control and quality certification problems in the main steam isola-tion valve closure system modification indicate that site engineer y
training, supervision, and performance should be improved.
The licensee has been developing engineer training programs, but imple-n mentation has not been accomplished.
(The licensee plans to conduct
job-specific training for site engineers prior to the July 1987 re-fueling outage.)
In summary, the engineering support organization was competent and s
staffed to support facility needs.
PRA support was excellent.
y
?
i ll '
Overall, appropriate attention to plant safety was assured by ex-D
%%
tensive reviews by diverse organizations.
However, poor coordina-
~
'
tion between site and corporate staffs was evident in the EQ and s
s t
.
.
O
_ - _ _ _, - _ - _ - - _ - _ - _ _ _
_ _. - _ - _ _ _ _. _ _ _ _ _ - _ _ - _ _ - - - - _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ - _ _.
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,-
fire protection programs. Also, problems with implementation of modifications indicated inadequate engineer training and/or ex-
perience.
2.
Conclusion Category 2 3.
Board Recommendations Licensee: Improve communication and coordination between the site and corporate staffs.
NRC:
None.
l
l l'
, _ _, _ _ _. _ _ _
_ _.. _ _ _ _ _ _ _. _.. _. _..... _...
.
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During the previous assessment period, this area was rated Category J
- .
'
?.
Further, c' declining trend was notedu based upon an increase
. ;
i in the number of incomplete and/or untimely submittals and the appearance of a decrease in overall management oversight in asturing i,
'
r,uali ty m ittals to the NRC. Better planning anc' attention to
-
,
regulatc,y deadlines was recommended.
'
'
'
'
$"
Thebasisofthecurrentappraisa(wat 'the licensee's support of N'
licensing actions that were eitheNcompleted or active during the ratf r4 perico. These activities consisted of amendment requests,
_
"j
,
-
'
' exem ion requests, responses ta generic letters, TMI Action Plan
-
itoms, Systematie ' Evaluation Program and Integrated Safety Assess-
.
,
, q. ?
maat Program (ISAM topics-, nrd their related licensing actions.
'
(
,
3 Licensing. activit'y during the current SALP period has remained very D
,
.)
high.
Forty-six (46) licensing actions were completed during this 13-month rating period compared to 50 licensing actions completed
^
, j.s during the previous 12 month rating period.
In addition to routine
.
'
actions, major activities that have been completed or are currently
'
ongoing incluce a fuel reload (Cycle 14), a steam generator tube i
sleeving program, initiation of tN Integrated Safety tissessment l
s
,
Program (ISAP) pilot program, enHremnertal qualification modifica-
'
t
'
tions, a schedular extension for implementing fire prctection re-
'
quirements (IC CFR 50.48), and modifications t) the DaePgency Core
,
Cooling systems to resolve recently discoverec: deficiencies in long term decay hett removal. A't.heugh 46 actions were completed during th*, rating period, 45 new Actions were added. At the completion of the' rating period, 53 licensing actions remairyi nctivo.
,
-
During this SALP period, the licensee has been heavily involved in
-
providing the results of the ISAP analyses for.the Haddam hek Plant.
The liconsee rubmitted both tae Haddam Neck probabilistic safety study (MS) and tneir proposed final ranking of licensing issues a.
in the IAAP. The PSS _was particularly valuable because it identi-fied semral areas of plant vulnerability (such as Motor Control Center Na. 5 syste'm interdependencies and certain small break loss-a e
of-coolant accident scenarios) which were not previously analyzed.
,
f Based upcn these vulnerabilfcies, the licenset took proapt interim
'
,
action to modify plant sys9ms and procedures. 1he licensee also
proposed future modi 4'ications to reduce these vulnerabilities.
'
,
A These actions demonstrate a strong management commhment to plant safety and a comprehensive effort to continue to upgrade safety at the Haddam Neck Plant,
'
'_
'
}
.,
_ _ _ _ _ _ _ - -
_ _ _ _ _
,
During the last rating period, several licensee submittals were untimely and/or inadequate for review.
It was concluded that man-agement oversight in assuring quality submittals did not appear to be at the level of previous periods.
During the first six months of the current rating period, the NRC observed an overall improve-ment in the quality and timeliness of the licensee's submittals.
However, the licensee did not sustain the improved performance dur-ing the remainder of the rating period.
During that period, the licensee requested seven waivers of compliance and/or emergency license amendments. The most notable subject area was the need to open manual containment isolation valves for routine plant opera-tions.
Manual isolation valve TS problems had been identified earlier but the licensee's submittals were not timely or comprehen-sive and addressed each necessary operation one at a time. Overall, this reflected a lack of comprehensive planning and timeliness in these submittals.
The licensee has usually exhibited an understanding of licensing issues and has generally employed a thorough and conservative ap-proach to address potential safety concerns. An example was the technical approach to resolution of issues concerning long term core cooling following a postulated break in the charging line. Another example was steam generator tube end cracking in all four steam generators.
These were examples of the licensee's clear understand-ing of the potential safety significance of each issue and consci-entious effort to comply with the regulations using conservative l
approaches. However, the approach to resolution of an additional l
emergency core cooling problem with a break in the core deluge sys-
!
tem was r.ot up to previous standards.
The licensee proposed re-solution by setting a flow control valve based upon calculation
,
l rather than an actual flow test, even though the flow rate was very sensitive to valve position and the margin for error was very small.
The staff could not conclude that the licensee employed a censerva-tive approach in resolving this issue, and required testing to prove l
the acceptability of the valve positioning.
l The licensee's response to NRC initiatives has generally been tech-nically thorough and usually tittely. When delays in providing in-formation have occurred, adequate justifications were usually pro-vided.
However, the licensee still has not been able to resolve several long-standing NRC concerns such as the NUREG-0737 technical specification upgrades and degraded grid voltage technical specifi-cations.
,
l In summary, licensee management and staff have demonstrated dedi-cated and competent involvement in safety at the Haddam Neck plant.
The licensee showed some improvement in the quality and timeliness of submittals but this was not sustained throughout the perio _- _ - - -
.
..
,
The submittal of information was generally timely and technically sound.
However, too many submittals continue to require the staff to request significant amounts of supporting technical information.
2.
. Conclusion
'
Category 2.
3.
Board Recommendations Licensee: Develop an overall approach for planning submittals to assure they are fully supported and submitted in a timely fashion.
NRC:
None.
.
l l
l l
_ _ _ _ _ _ _ _ _ _ _ _ _ _. _... _ _ _ _ _ _ _ _.. _ _ _
. _ _ _ _ _ _ _ _ _ _ _
____ _ _
.
[
,
G.
Refueling and Outage Management (240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, 7%)
1.
. Analysis During the previous SALP, this area was rated Category 2.
NRC con-cerns centered on inadequate preplanning and preparation of modifi-cations and the reactive mode of management caused by the need to simultaneously plan, review, and implement these changes.
At the beginning of the current assessment period, the 1986 outage was still in progress.
It was completed two months later.
Inade-quate preplanning of jobs for this outage reflected an inadequate corporate / site interface and challenged the licensee's plant man-agement and supervisory capabilities.
Extraordinary plant staff efforts generally overcame the negative impact of inadequate pre-planning.
Refueling outage coordination amployed comouter-assisted critical path scheduling and management coordinators on'each shift.
Project schedules received wide distribution and were reviewed twice each day at plant status meetings.
An indication of the adverse effect of the high outage workload was the identification, after plant start-up, that a degraded steam generator U-tube had not been plugged.
In 1987, licensee management initiated efforts to improve preplan-ning of the refueling outage scheduled to begin in July 1987.
Re-gular outage planning meetings were widely attended by organizations involved in outage planning, scheduling, and implementation. NRC observers particularly noted the development of pre-outage activity schedules which track the development and planning of outage jobs using the same detail and critical path tracking used during outages.
An unplanned shutdown and 15 day maintaance outage occurred in July 1986 because of a charging pump failure and the subsequent plugging of previously identified defects in many steam generator U-tubes.
This identification was due to the application of improved surveil-lance capabilities. The licensee implemented work activities in accordance with a preplanned shutdown work list which was supple-mented by the steam generator repair activities schedule.
The main inspection effort during this SALP period occurred during the last two months of the 1986 refueling outage. The assessment of that outage, as described in the previous SALP, is unchanged.
While positive steps have been made to improve this area, results of these efforts will not be demonstrated until they are challenged during the next refueling outag. _ _ - _ _ - _.
-
c 2.
Conclusion Category 2.
3.
Board Recommendations None.
_ ________
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_
,
i*
31-
!
H.
Radiolog'ical Controls - (507 hours0.00587 days <br />0.141 hours <br />8.382936e-4 weeks <br />1.929135e-4 months <br />,14%)-
1.
Analysis The last SALP rated this area as Category 2.
Concerns included:
recurrence of minor, self-identified violations; weak job planning, faulty' procedure adherence and corrective actions, and use of poor
~
t
judgement; and tracking of exposures without effective control.
A radioactive drum was compacted in violation of procedures, with unnecessary worker exposure and work area contamination. Also, a worker exceeded his assigned exposure by about a factor of two while staying unnecessarily in a high radiation area to resolve a problem with mismatched couplings.
Exposures were consistently higher than licensee projections.
Proposed corrective actions were not effec-tively supported by management. Comprehensive licensee review and upgrade of the ALARA program was recommended.
The radiological controls program and organization are acceptable.
Lines of authority are clear.
The staff is professionally capable.
During this SALP period, external radiation control, dosimetry, and respiratory protection were generally. good. The radiological con-trols staff provided many suggestions for improvement. Management wasLgenerally responsive. A noteworthy example was implementation-of a zone concept for control of radiation work areas, with each zone supervised by a health physics foreman, improving supervisory oversight.
Escalated enforcement action was taken for the exposure of a steam generator worker to a quarterly total of 3.3 rem (3 rem federal limit)-.
Station health physics supervision had not been overseeing Lor monitoring this very high radiation field work.
Station manage-ment had not observed and evaluated the activity. This event re-flected unacceptable program implementation and management. NRC review also concluded that the licensee had treated health physics events as individual failures to follow procedures without address-ing the related supervisory and program implementation inadequacies.
Evaluation of the effectiveness of the licensee's corrective actions awaits observation of refueling outage activities.
Radiation protection procedure inadequacies remain an NRC concern.
i Survey and radioactivity counting procedures are ambiguous. Many
]
procedures are difficult to understand.
Key concepts are omitted
or poorly defined. There are multiple technical and typographical errors. A typical inadequacy was the specifying of undefined acti-vities (e.g., intermittent surveys, continuous coverage) in proce-dures.
These were not well understood and contributed to health physics events including the overexposure identified in the preced-ing paragraph. This demonstrates lack of attention to detail in procedure development and inadequate initial and periodic procedure review.
___________________ - _
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-_.
_ _ - _ _ _ _ _ - _
____
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"
.
[,
-32 l
- Deviation from procedures also remains an NRC concern. An effective l
- means of assuring technician familiarity with procedures affecting their work areas, and with associated changes., has not been evident.
The ineffectiveness of the existing briefings by supervisors and l
'. individual certifications of knowledge and understanding was shown by inadequate control of significant radiological operations and of high radiation area entries.
Despite the management support evident in the raising of the value of preventing a man-rem to $20,000, ALARA weaknesses continue. The ALARA staff consists of a coordinator and an assistant. Though qualified, they were overwhelmed by the number and scope of needed ALARA initiatives.
Computers were available, but a lack of software necessitated use of less efficient manusi methods.
For example, ALARA personnel were observed to be manually totalling exposure data to the detriment of their other duties.
Radiological controls per-
-sonnel appeared ineffective in limiting access, stay times, and nonproductive entries. Job completion was often delayed, with im-proper fit of components a typical problem.
Packages for onsite ALARA review were typically incomplete and so late that there was insufficient review time.
Exposure reduction. initiatives focused.
on shielding, decontamination, job cancellation, and other technical aspects. Weaknesses'in access and stay time control were generally not addressed. ALARA goals established by the corporate staff were not based on specific job analyses, and were therefore regularly exceeded.
The 1700 man-rem exposure for 1986 is considered to be
- unnecessarily and extremely high. While the ALARA goal for steam generator work in July 1986 was achieved and exposure goals are normally met for non-outage work, the problems identified show that ALARA controls have not been properly effective.
A December 1986 licensee ALARA appraisal attributed unnecessary radiation exposure to several of the above weaknesses.
Corrective actions were not formulated during the SALP period.
Improved pre-parations for the 1987 refueling outage have, however, been evident.
Regular planning sessions have included ALARA personnel. Several iterations of ALARA goals for outage projects have been evident.
Daily management planning meetings have incorporated increased at-tention to radiation exposure goals.
Long-standing program defi-ciencies have been addressed by clear assigerent of responsibility for exposure controls and by more stringent management insistence upon procedure adherence. The onsite health physicist position was filled by an individual who appears to be changing the job function from a consulting role to one of active participation in ongoing activities. Also, a new staff position was addea under the Station Services Superintendent. That position, although it does not directly involve the radiation protection staff, should free the station services superintendent for more attention to health physics matters.
These steps show management support, but overall assessment of their effectiveness must await evaluation of the forthcoming refueling outage.
- _ = _ - _
_ _ _ __-_-
t
j The c; porate
'ce provided the site with standardized health physics prot.euures.
These,.though different from the ones used on site, were generally very well written and provided valuable tech-nical references. Monthly corporate audits of site activities were performed. The NRC concluded that these audits were not of suffi-cient depth and breadth to effect significant program improvements.
'Overall, corporate radiation protection staff input was assessed as not having a significant impact on site radiation protection.
INP0 accreditat%n of health physics technician training is sched-uled for 1987 ihis training has been upgraded in scope and depth.
Lesson plans have been improved.
More time is spent in training.
Effectiveness is better measured by increased use of exams.
These are substantial program improvements, but newness prevented assess-ment of their effectiveness.
Facilities and equipment were generally adequate.
Concerns included irregular timing of quality control (QC) checks on instruments and QC checks being performed without a clear understanding of the sig-nificance of the results.
For example, control charts can be used to plot results of suc.cessive tests, with the acceptance band clearly denoted.
Chart examination illustrates both the current performance and the trend.
Such control charts were not used at Haddam Neck, and NRC inspection found a lack of technician awareness of the associated performance characteristics. Also, although access to computers was available, software development has lagged.
The resultant need to perform functions manually was assessed as prone to error and an inefficient use of licensee staff time.
Effluent controls are the responsibility of the chemistry organiza-tion.
Extensive laboratory training facilities including plant specific, state of the art equipment have been completed.
Based on participation in the new chemistry program and inspector discus-sions with technicians, the NRC concluded that the chemistry staff was acceptably qualified.
INPO accreditation of chemistry techni-cians is scheduled for 1987.
The offsite dose calculation manual was generally implemented as required by the Radiological Effluent Technical Specifications (RETS).
However, the licensee did not perform a pre-implementation audit of the RETS. Onsite licensee QA review of RETS surveillance implementation was scheduled for six months after program implemen-tation.
Four months after implementation, the chemistry organiza-tion identified two surveillance which were not addressed in pro-cedures and initiated procedure coverage.
No additional licensee review was implemented.
Later, the NRC identified two violations of the RETS. One involved the procedures not prescribing all of the sampling and analysis frequencies for continuous liquid releases.
The second involved daily grab samples not being collected and weekly analyses not being performed for continuous releases from
_ _ - _ _ _ _ _ _ _ _ _ _ _ _._. _
.
__. _--
.
,
steam generator blowdown and service water effluent.
The NRC also identified record management weaknesses.
These included undated offsite dose calculations which did not demonstrate tsmp?iance with the 31-day frequency requirement and unavailability of a gamma re-lease record for a batch liquid release.
Overall, RETS inplementa-tion reflected a need for better site and corporate interaction and for pre-implementation validation of new programs and procedures.
Radwaste was generally well controlled.
Processing systems wera well maintained. Outstanding maintenance actions were less than two months old.
Licensee personnel showed good understanding of system capabilities and alarm responses.
Minor weaknesses were evident in violations related to characterization and documentation of radionuclides in radwaste shipments.
Previous NRC appraisals had highlighted inadequacies in QA for waste packaging and shipment; corrective actions had not prevented subsequent violations.
Cor-rective actions now appear effective, in that no subsequent problems have been identified. Also, after the SALP period, NRC inspection found the programmatic deficiencies to be corrected.
In summary, the basically sound radiation controls performance has been marred by an above limit personnel exporure, by weaknesses in exposure controls, by procedure inadequacies, by failures to follow procedures, by ALARA weaknesses, by a lack of effective corporate radiation protection staff input, and by inadequacies in RETS im-plementation.
Recent site management attention to problems has been obvious, but the results will not be tangible until the refueling outage occurs.
2.
Conclusion:
Category 2 3.
Board Recommendation Licensee: Assure effective site and corporate coordination and pre-planning of radiological controls activities.
NRC:
Perform a special inspection of refueling outage radiation controls.
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Emergency Preparedney (201 hours0.00233 days <br />0.0558 hours <br />3.323413e-4 weeks <br />7.64805e-5 months <br />, 6%)
1.
Analysis During the previous assessment period the licensee was rated Cate-gory 2 in this area, principally because of weaknesses noted in communications between emergency response facilities and the number of weaknesses noted during the annual exercise.
During the current assessment period there were two region-based inspections. One inspection included NRC observation of the licen-see's April 25-26, 1986 annual full participation emergency pre-paredness exercise. Another inspection covered changes to the Emergency Preparedness (EP) Program; shift staffing and augmentation; knowledge and performance of duties; dose calculation and assessment; and licensee audits.
During the 1986 emergency exercise the NRC observation team identi-fied several strengths and weaknesses in the licensee's EP program.
The overall assessment was that performance provided reasonable assurance that the public could be protected.
Specific strengths were noted in the areas of accident mitigation, communications with the state, classification of plant conditions, protective action recommendations, rumor control, press releases, and strong technical management in the corporate Emergency Operations Center (EOC).
Weaknesses identified by the previous exercise as requiring correc-tive action were adequately demonstrated and closed.
However, several new items were identified as potential weaknesses.
The more significant of these include:
Control room personnel interchanged ALERT and SITE AREA
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EMERGENCY terminology during communications; Site evacuation was ordered without consideration of the tor-
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nado watch which was in effect;
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Director of Station Emergency Operations (DSEO) did not inform the Manager of Control Room Operations that he was assuming DSE0 duties; and Communication between the site and corporate E0Cs was weak,
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in that the corporate E00 was not clearly informed of the basis for the SITE AREA EMERGENCY declaration.
Communication of information between remote emergency operating facilities has been a long standing weakness.
The above noted ex-ercise observations indicated that communication problems were still
,
evident. Also, NRC observed slow implementation of control room
data transmission functions during a March 1987 emergency staff augmentation drill.
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!h NRC review'of-EP implemen_ ting procedures (EPIPs) revealed that there
was no recognition'of an~ interface between the licensee's DSE0 and an NRC Director _ of Site Operations (DS0)/ Site Team Leader, if an NRC response team was on-site.
Specifically, the need to keep the NRC-DSO informed of any notifications / protective action recommenda-tions being transmitted to the State was'not'specified. The.licen-see.was responsive to this concern and agreed to make necessary 1 procedure changes and. stress this issue during. training. Another NRC concern was the effectiveness of theLemergency staff augmenta ~
tion plan'..The licensee's Emergency Plan extends the allowable-staff augmentation time to one hour rather than the recommended 30 minutes. This capability had not been demonstrated by unannounced call-in exercises. This concern was reemphasized on March 24, 1987 when the licensee conducted an unannounced drill and three responders-did not reach the' site.within one hour.
Strong management response was noted in emphasizing both personal and supervisoi accountability
- for'the performance of scheduled responders. The licensee plans
'to continue to randomly exercise the augmentation plan.
The licensee's emergency response facilities are generally well-designed and complete with the exception of the installation of a safety parameter display system-(SPDS) and the monitoring of certain process variables.. The system improvements await installation of a new main frame computer which is scheduled'for the July 1987 out-
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age.-
Final implementation and training for SDPS is: scheduled for the spring of 1988. This schedule has been accepted by NRC.
Upon completion, the licensee facilities will be fully assessed in an NRC Emergency Response Facilities Appraisal.
The' licensee's corporate organization performs audits and appraisals of the emergency plan, EPIPs and corporate emergency procedures.
These audits cover all the major elements of the EP program at least once every 12 months. The audits are thorough and detailed, and provide assurance that weaknesses or potential weaknesses will be identified. The audit review and distribution process is extensive and adequate administrative controls exist for documentation and follow-up of corrective actions.
- The Emergency Preparedness Staff at Haddam Neck is adequate, con-sisting of an onsite Emergency Preparedness Coordinator and Assist-ant Emergency Preparedness Coordinator. Additional assistancr. is available from the Supervisor Emergency Preparedness at the Corpor-ate Headquarters in Berlin, Connecticut. The licensee continues to maintain an excellent working relationship with the State of Connecticut and local governmental agencies as evidenced by the continuing cooperation demonstrated during exercise.---_-. _ _ _ -
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Overall, the licensee maintains an EP program capable of providing adequate protective measures in the event of an emergency. Corpor-ate involvement in EP is evident and slow but steady improvements have been observed. The good working relationship with the State of Connecticut is considered a strength.
Exercise performance is satisfactory, but better communications between E0Cs and the control i
room, improved emergency staff augmentation performance, and com-pletion of outstanding response facility hardware upgrades will increase the'overall effectiveness of the licensee's emergency response functions.
2.
Conclusion Category 2.
3.
Board Recommendation None
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Security and Safeguards (188 hours0.00218 days <br />0.0522 hours <br />3.108466e-4 weeks <br />7.1534e-5 months <br />, 5%)
1.
Analysi s-During the previous SALP period, no regulatory concerns were iden-tified and.the licensee's performance was assessed as Category 1.
.
-The NRC has not cited any violations of its requirements during four consecutive rating periods (1980 to present).
- The licensee has maintained this high level of security program effectiveness for a period of almost seven years.
NRC attributes this to:
strong management involvement and support for the program in-
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cluding effective licensee oversight of the contract security force; a comprehensive security program based on NRC performance ob-
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jectives rather than minimum regulatory requirements, and up-grading of that program to meet changing needs on a regular basis; utilizing state-of-the-art systems and equipment and maintain-
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ing systems and equipment in good working order by providing the necessary technical' oversight;
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establishing and maintaining a well-trained, aggressive and
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highly notivated security organization that emphasizes personal accountability; and compensation for a high security force turnover rate by estab-
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lishing and maintaining a' highly effective training program that includes feedback derived from personnel performance,
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audits and surveillance, and program upgrades.
Corporate support for the program was demonstrated by the licensee's active participation in various nuclear industry professional groups and associations and by providing the necessary resources to upgrade
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the program on a regular basis. An example of this support was the i
licensee's development of a read and sign indoctrination training program for NRC inspectors. This program significantly improved the licensee's capability to provide prompt, unfettered NRC access.
To ensure the continuing effectiveness of the security program, the licensee utilizes comprehensive corporate auditing and site-based
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self-assessment programs that are carried out by personnel with nuclear plant security expertise. The corporate audit program is p
conducted on an unannounced basis and includes periodic reviews of selected aspects of the security program throughout the year. All
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aspects of the program are reviewed by the end of the audit year.
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Audit reports were distributed to responsible senior management per-sonnel for review. Corrective actions on identified deficiencies were-prompt and effective in all cases and followup was conducted to ensure that corrective actions were effective.
The site security organization also performed continuing self-assessments of the pro-gram using NRC inspection criteria and program performance objec-tives..These assessments were conducted independently by propri-etary supervisors or, in some cases, by contract security force supervisors with proprietary oversight.
Using these techniques, the licensee was able to identify and correct weaknesses and poten-tial violations before they adversely affected the program.
Ex-amples are listed below.
Four security event reports were submitted in accordance with 10 CFR 73.71. One event involved a brief system failure as a result of a component malfunction.
The remaining events involved the lic-ensee's identification and repair of two vital area barrier weak-nesses and of an isolated personnel error which resulted in an un-secured vital area door.
The licensee's response to these events was prompt and thorough.
In the case of the barrier weaknesses, those problems were identified during scheduled security plan self-assessment efforts and follow-up on an NRC Information Notice.
Each event was appropriately handled and compensatory measures were promptly initiated, when required.
Event reports were clear, con-cise and submitted to NRC in a timely fashion. The small number of reportable events is further evidence of the quality of the lic-ensee's program.
The licensee's security organization was adequately staffed with well qualified personnel.
Both proprietary and contractor super-visors were well trained, professional and provided effective super-vision. A noteworthy strength in the organization was the licen-see's policy of requiring security force members to qualify by ex-amination in all positions, up to Alarm Station Operator, before being eligible for promotion to the rank of Sergeant. This policy ensured a detailed knowledge of security program requirements, sys-tems and equipment prior to consideration for a supervisory position.
This appeared to significantly enhance program implementation and quality. All members of the force were well trained, knowledgeable of +,he program, and highly motivated.
This was consistently demon-strated in interviews and records reviews conducted durir.g NRC in-spections and again demor.strates the licensee's commitment to a quality program.
Facilities were clean and well maintained.
Sufficient space was l
allocated for the administrative and operational needs of the pro-gram.
Records were also well maintained and readily retrievable.
I Record repositories were found to be in accordance with NRC require-ments and properly secured.
1.
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l Training facilities are well designed and provide for the efficient
. administration of the training program. The training was admini-stered by qualified,' full-time instructors. ' Lesson plans and.in-structional aids were maintained current utilizing. feedback from day-to-day operations and the-audit and self-assessment programs.
The training instructors also. played a key role in evaluating the effectiveness of-the security force. drill program and in assessing
individual and team proficiency levels during these drills. A total e
of 196 drills and 75 individual performance evaluations were con-ducted during 1986. These drills served to reinforce the training and qualification program and increase the performance level and proficiency of the security force.
Two revisions to the Security Plan were submitted-under the provi-
'
sions of 10 CFR 50.54(p) and the licensee also responded to the Amendment to'10 CFR 73.55, codified by NRC on August 4, 1986.
The
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-licensee's staff responsible for these actions had a thorough
)
knowledge and understanding of NRC security program objectives..
and ensured that plans werel current and that changes were. properly coordinated, when required. Security Plan changes were coordinated with regional safeguards licensing personnel to ensure that a clear understanding of.each change existed. These changes were usually.
of high quality, indicating appropriate staffing and in-depth man-agement review. With respect to the revision currently under NRC review, some changes were not adequately reflected in the text of the plan.
In summary, the licensee continued to maintain a highly effective and proactive security program. This achievement reflects favorably upon the managers and supervisors,at both corporate and site levels, and on members of the security force who have aggressively pursued excellence in the discharge of their responsibilities. Management attention to and support of the program were clearly evident from the high degree of. success which was achieved.
2.
Conclusion Category 1.
3.
Board Recommendation None.
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K Training and Qualification Effectiveness 1.
Analysis Training and Qualification Effectiveness is an evaluation criterion for each functional area. During this SALP, it also is being separately addressed as a synopsis of the assessments in the other areas.
Training effectiveness has been measured primarily by the observed performance of licensee personnel and, to a lesser degree, through program review.
Training effectiveness was rated Category 2 during the previous SALP period.
NRC concerns included recurrent weaknesses in ALARA, modi-fication control and fire protection. During this SALP period, training contributed to the reduction in fire barrier related re-portable events. Weaknesses in exposure control and modifications were again apparent, however.
Management commitment to quality training was evident from the ex-cellent fa'cilities and staff which have been developed.
During this SALP period, the operator license programs received INP0 accredita-tion and a new, plant specific simulator was put into operation.
In addition, senior operations personnel were promoted to the training staff, improving training background in operations. The training staff includes more than 20 instructors, most of whom hold NRC licenses or instructor certifications.
However, many of these
. instructors are not licensee employees. The licensee has made an effort to reduce the number of contractors in the training organi-zation, but the large number remaining provides a high potential for an unanticipated reduction of plant specifk expertise.
However, no associated impact on operator performance or training effective-ness was identified during this SALP period.
Initial operator training continued to be a program strength with 14 of 15 successful license candidates during this SALP period.
Significant improvement was noted in the operator requalification program, although administrative controls were not completely im-plemented.
Technical staff training was also improved in most functional areas during this period. This training includes academic instruction in areas such as mathematics and science as necessary for job-related tasks, and practical applications in the laboratory and jobsite locations. While the training organization is located off-site, sufficient training staff works at the site to provide on-the-job training and operational feedback to the training programs.
The technical training programs are scheduled for INPO accreditation in 1987.
Personnel interviews have indicated generally positive evaluations of the quality and applicability of the new training initiatives.
On the other hand, operator and maintenance errors
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caused three plant trips during this SALP period and there continued to be a high proportion of reported events involving personnel error.
In addition, weaknesses in chemistry department training related to implementation of the Radiological Effluent Technical Specifica-tions contributed to several NRC and licensee identified problems in that area.
These factors and the recurrent instances of proce-dural compliance and documentation inadequacies noted in the main-tenance, surveillance, and radiation protection areas indicate that training in these areas has not yet been properly effective.
The licensee has not yet implemented formal training programs ad-dressing concerns about modification control and testing.
This lack of training, particularly for relatively inexperienced plant staff members, has contributed to recurrent problems such as inadequate post modification testing of the Appendix R shutdown instrumentation and of a replacement containment isolation valve.
The General employee training (GET) program adequately addressed orientation, radiation protection, security, emergency planning, safety and assurance of quality. GET content is directed by a steering committee made up of station managers who determine program emphasis based on station performance goals.
During this SALP period, GET emphasized the importance of fire barrier control.
No further events occurred in this area and NRC noted appropriate con-cern for fire barriers during discussions with personnel in the field.
The licensee's extensive training program for the plant security staff was judged to be a significant factor in the continuing ex-cellence of the security organization.
In summary, the licensee's significant commitment to training is evident in the development of extensive training facilities and competent management and staff. Operator license training programs were effective and the security staff training was excellent. With the exception of engineer training, improvement was noted in the technical training programs.
However, training did not prevent recurrent problems with personnel errors and radiction exposure control.
2.
Conclusion Category 2 3.
Board Recommendation None l'
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. Assurance of Quality 1.
Analysis Management involvement in assuring quality is an evaluation cri-terion for each functional area.
Quality assurance (QA) also is an integral part of each functional area. This Assurance of Quality area is a synopsis of the applicable aspects of other areas, in-cluding worker ar,d supervisor performance, management oversight, and safety review committee activities.
The last SALP rated this area as Category 2, with the major concern being for the effective-ness of self-evaluation and resolution of problems.
It was recom-mended that the associated licensee programs be reevaluated.
During the current SALP period, licensee personnel were assessed as technically competent and as focused on doing things proptrly the first time. Most activities were carefully conducted in ac-cordance with relevant directives. -The necessity for careful at-tention to detail was acknowledged.
In contrast, a willingness to deviate from procedural requirements was evident from the analyses of the preceding functional areas. Also, the total of 20 violations and 17 personnel error related events (including three plant trips)
showed multiple inadequacies in adherence to requirements.
The number and nature of these occurrences show substantial room for improvement in performance and supervisory overview.
Breakdowns in the overall assurance of quality were evident in the above limit exposure of a steam generator worker and in the rest of the violations of NRC requirements.
For the overexposure, it was particularly relevant that neither senior health physics nor senior station managers had provided evaluation / overview of the activity, which had been ongoing for several weeks when the over-exposure occurred.
First line supervisors generally provided oversight of activities and were knowledgeable of associated design and administrative re-quirements.
They were often observed to be providing guidance at work sites.
Operating shift supervisors exhibited thorough knowi-edge of plant activities. Daily first line supervisor meetings were held to coordinate work activities, and this new initiative was found to be an effective means of controlling mutual interference.
Department supervisors also were observed to be in the plant fre-quently.
They were found knowledgeable of their departments' acti-vities and of significant plant problems.
Daily meetings to discuss planned activities, review problems, and assign corrective action responsibilities were considered effective.
Licensee tracking sys-tems were found effective in closing items, except that the large number of NRC items open for a prolonged period indicates a lack of aggressive licensee follow-up of such items.
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a The unit and plant superintendents made frequent control room and plant tours.
Formal weekly inspections by the superintendents and department supervisors were made, and improved housekeeping was one result.
Senior plant staff members acted as management duty officers during operations and outages.
In addition, management representatives were assigned for full time coverage of outages. These measures and the daily staff meetings were generally effective in assuring proper activity control.
Station management was noted to be ac-tively involved in the daily meetings, with action items and due dates being generated as a result.
Rapid identification of problems was achieved by the Plant Incident Report (PIR) system.
There was a low threshold for PIR initiation, and over 200 PIRs were written in 1986. The PIRs were an excellent l
tool for site managers, who maintained awareness of the associated root causes and corrective actions.
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QA routinely provided inputs to plant managers on the effectiveness I
of activities.
There were frequent quality control (QC) inspections of the proper completion of safety-significant jobs.
QA/QC person-nel were appropriately qualified and certified.
QA surveillance of in process activities was increased, with about 70 separate evolutions observed.
Findings were provided to depart-ment supervisors, and corrective actions were verified during sub-sequent surveillance.
The QA aedit system functioned acceptably.
Schedules were followed.
Checklists were well organized and comprehensive.
Corrective action timeliness and thoroughness remains a concern, however.
Procedural deficiencies were the subject of audit findings in 1981, 1983, 1985, and 1986, clearly indicating ineffective correction of this generic problem.
The Plant Operations Review Committee (PORC) and the Nuclear Review Board (offsite committee) provide quality oversight of safety-related activities.
NRC observed frank, open, and knowledgeable discussions of issues and a sound approach to safety was clearly demonstrated.
The contribution of these committees was shown both in the routine review of plant documents and changes, and in assess-ing the safety significance of two safety system design errors which were identified during this SALP period.
The committees were in-strumental in development of corrective actions with appropriate attention to continued plant safety.
Nuclear Safety Engineering (NSE), the independent safety engineering group which is part of the corporate staff, was active in its coverage at Haddam Neck. This onsite group had ready access to the
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- plant staff, equipment, and records. NSE' assessed plant safety programs and evaluated plant operating experiences through reviews
'of, procedures and data-including independent reviews of the resolu-tion of ~ome' Licensee Event Reports (LERs)'and PIRs.
In addition, s
NSE performed special~ reviews and evaluations as requested by plant management.- An example is an ongoing NSE review of industry pro-cedures.
!
In summary, assurance of quality was generally satisfactory. Good-
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performance was evident in most aspects.of performance.
But there was unacceptable control over the high radiation work involved in steam generator entries, an ineffective ALARA program, a high number of-plant trips, a.high number of_vio.lations.of NRC requirements, and a_high' number of personnel errors resulting in plant transients and events.
Sound recent management initiatives were evident in outage planning and in initiation of action items and schedular tracking 'on. problems -identified during daily activities. Overall, it' appears that recent positive initiatives have, so far, had less affect on performance than.long-standing practices'and attitudes.
2. - - Conclusion-
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Category 2.
3.
Board Recommendation None
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V.
SUPPORTING DATA AND SUMMARIES A.
Investigation and Allegation Review None B.
Escalated Enforcement Actions 1.
Civil Penalties A $50,000 civil penalty was issued on December 10, 1986 for an occupational exposure exceeding 10 CFR 20 quarterly limits.
2.
Orders None 3.
Confirmatory Action Letters None C.
Management Conferences 1.
On June 18, 1986, an enforcement conference was held at the NRC Region I office to discuss violations and program weaknesses related to the radioactive waste transportation area.
2.
On April 7, 1986, a management meeting was held at the Northeast Utilities Corporate Office to discuss the results of the independent review of the design change control program required by NRC Order dated December 13, 1984.
3.
On September 3, 1986, an enforcement conference was held at the NRC Region I office to discuss an above limit occupational whole body exposure and several violations of NRC fire protection requirements.
D.
Review of Licensee Event Reports (LERs)
1.
LERs Reviewed LER No. 86-10 to 87-03
,
2.
Tabular Listing a.
Personnel Errors
b.
Design / Man./Const./ Install
c.
External Cause
)
d.
Defective Procedure
)
e.
Component Failure
x.
Other
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Total
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A tabulation of LERs by functional area and an LER synopsis are attached as Table 4.
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3.
Causal Analysis (Review Period 3/1/84 - 3/31/87)
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i Four sets of common mode events were identfied:
a.
LERs'86-21, 22, 23, 24, 30, 32 and 87-02 reported age-related nuclear instrumentation anomalies.
Five of these resulted in turbine load runback actuations.
b.
LERs 85-17, 85-29, 86-13, and 86-48 reported engineered safe-guards systems design / analysis errors.
c.
LERs 84-10, 86-04, 86-33 and 86-46 reported problems with operability of the low temperature overpressure protection system.
d.
LERs 86-43 and 87-03 reported problems with the operability of the wide range plant stack monitor.
Other notable trends were the continued high percentage of personnel errors and component failures.
E.
Summary of Licensing Activities 1.
Schedular Extensions Granted August 25, 1986 Schedular Extension for 10 CFR 50.48, Switchgear Room Modifications December 15, 1986 Schedular Extension to Commitments for Halon System Modifications in the Switch-gear Room 2.
Reliefs Granted May 12, 1986 Relief from Inservice Inspection Require-ments from the 1980 Edition of the ASME Code 3.
Exemptions Granted April 28, 1986 Exemption from General Design Criteria 35 - Small Break LOCA Single Failure Requirements l
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4.
License Amendments Issued April 14,1986 Amendment 73 - Primary System Leakage April 14,1986 Amendment 74 - Cycle 14 Reload April 29,1986 Amendment 75 - Fire Detection and Spray Systems
- May 14, 1986 Amendment 76 - Steam Generator Tube Plugging
- July 11, 1986 Amendment 77 - High Pressure Recirculation for Emergency Core Cooling Systems July 14, 1986 Amendment 78 - Steam Generator Tube Sleeving August 6, 1986 Amendment 79 - Administrative and Reporting Requirements August 7, 1986 Amendment 80 - Minimum Shift Crew Composi-tion August 18, 1986 Amendment 81 - Control Room Fire Detection System
- September 3, 1986 Amendment 82 - Inservice Inspection of Steam Generator Tubes September 9, 1986 Amendment 83 - Fire Protection Audits September 18, 1986 Amendment 84 - Quadrant Power Tilt Ratio September 29, 1986 Amendment 85 - Monthly Operating Reports
- October 30, 1986 Amendment 86 - Manual Containment Isolation Valves November 12, 1986 Amendment 87 - Inservice Inspection of Reactor Coolant Pump Flywheels December 24, 1986 Amendment 88 - RHR Flow Control Valve 796 in Emergency Core Cooling System February 9, 1987 Amendment 89 - Charging Pump Technical Specifications
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- March 11, 1987 Amendment 90 - Manual Containment Isolation Valves (Neutron Shield Tank)
" Waiver of Technical Specifications compliance granted.
5.
Emergency Technical Specifications Issued May 14, 1986 Amendment 76 - Steam Generator Tube Plugging December 24, 1986 Amendment 88 - RHR Flow Control Valve 796 in Emergency Core Cooling System 6.
Orders Issued July 2,1986 Revision to TMI Confirmatory Order Related to SPDS/DCRDR Schedules F.
Description of Plant Activities (3/1/86 - 3/31/87)
At the beginning of this SALP period (March 1,1986), the facility had been shutdown for a refueling and maintenance outage since January 4, 1986. Outage activities included the core XIV reload, steam generator eddy current testing and tube plugging, containment leak rate testing, backfit modifications and plant equipment upgrades, and secondary system overhaul / repair work. On March 2, the licensee safely recovered the fuel element which was dropped onto the core February 26,.1986, during reactor disassembly for fuel offload. On April 1, 1986, the licensee reported an error in the existing plant safety analysis for small break loss of coolant accidents (LOCA).
Certain small breaks at one reactor coolant system location could preclude the proper operation of emergency core cooling systems (ECCS) in the containment recirculation mode.
The lic-ensee developed a set of administrative and procedural controls to mini-mize this event pending ECCS modifications scheduled for a future outage.
NRC approved the licensee's program including a temporary exemption from the single failure criterion on April 28, 1986.
l Plant heatup began on April 27, 1986, and low power physics testing was conducted satisfactorily on May 6-8, 1986. On May 7, the licensee dis-covered that one defective steam generator (SG) U-tube in SG #2 had not been plugged as required by Technical Specifications (TS).
The licensee isolated SG #2 and reactor coolant system (RCS) loop #2 and requested TS relief to allow cycle XIV operation with the 55% through-wall defect unplugged. NRC Licensing granted this relief on May 8,1986. While still in 3-loop operation on May 8, the plant automatically tripped from j
10 percent power.
The blocked main steam isolation valve (MSIV) trip signal, which existed because of the closed MSIV on SG #2, automatically ur. blocked when power was raised above the 10% trip reset point during a main turbine balancing evolution.
The trip was due to operator in-
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W attention to the existing blocked trip signal and its automatic unblock setpoint. While the plant was shutdown, SG #2 and RCS loop #2 were un-isolated since TS relief had been granted to allow operation of SG #2.
The reactor was restarted on May 8 and the unit was phased to the grid on May 10.
Full power was reached on May 22, 1986.
During this assessment period, spurious fluctuations in nuclear instru-mentation (NIS) channels caused several inadvertent turbine load runback actuations.
The licensee determined that component aging within the NIS was the primary cause of these fluctuations.
In addition to repairing individual-failed components, a project to replace the NIS with state of the art equipment has been initiated.
On June 4, plant operators manually tripped the plant at full power due to a loss of main feedwater (rapidly decreasing SG levels) caused by a failed closed heater drain tank level control valve (LCV).
Following repairs to the LCV on June 5, the plant operated at full power until June 17, when the same LCV again failed shut and operators again manually tripped the plant in anticipation of an automatic trip. This time the LCV was modified to prevent the separation of the valve stem and operator, which caused both the June 4 and 17 trips.
Main Steam Isolation Valve (MSIV) stroke testing following the June 17 trip revealed that the MSIV air operating accumulator did not have adequate capacity to close all four MSIVs without non-safety-related make-up air. The reactor was re-started, but power 'peration was delayed until MSIV air system modifica-tions were completed on June 22. During this period, an automatic reat-tor trip from.zero power occurred on June 19 while troubleshooting a failed NIS power range drawer. Operators and technicians failed to re-cognize the consequences of re-energizing the shorted NIS drawer. A second trip occurred during plant startup on June 22. The plant had been placed in a 3-loop operating condition to measure RCS flow in this operating mode. A reactor protection system (RPS) loop low flow relay contact had been obstructed by a small piece of ceiling tile, creating i
an unnannunciated reactor loop low flow trip signal.
Two low flow trip i
signals are required to trip the reactor between 10% and 74% power. A second trip signal was actuated and annunciated by the idle loop in the
3-loop operating mode. When the low flow trip signal was automatically
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unblocked by design at 10% power, the reactor tripped. The ceiling tile most likely entered the RPS relay as a result of control room work during the 1986 refueling outage. Upon identification and correction of the cause of the trip, plant operation resumed on June 24.
Full power was reached on June 26 following a controlled shutdown on June 25 to recover the idle RCS loop.
The plant operated at full power until July 11, when a load reduction was initiated due to failure of the "A" charging pump. The pump, which failed on July 8, could not be repaired prior to expiration of the TS limiting condition for operation. Therefore, the unit was placed in a hot shutdown condition. On July 15, the licensee placed the unit in cold shutdown to accomplish steam generator (SG) U-tube repairs. This 3 week l
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YIi unplanned shutdown resulted from an evaluation of SG eddy-current test
'
%m data.(from the 1986' refueling / maintenance outage), which indicated pre-
viously unidentified SG iube cracks in the rolled area of the U-tubes.
'
3'
During the' mini-outage, on' July 23, a contractor SG worker was exposed
.to 110% of his allowable quarterly radiation exposure limit..Following
.
the. mini-outage, full power was reached on August 6.
On August 30, operators manually tripped the plant after a' failed-open feedwater regu--
' latin ~g valve (FRV) resulted in rapidly increasing SG 1evels.
Power operation resumed on September 3, 1986, upon completion of FRV repairs.
"
Full power operation continued'-(except for brief load reductions for
ro'utine tests and maintenance) until November 30, 1986', when an automatic trip occurred when a technician shorted the SG feedwater control system n
while repairing an energized circuit.
Increased RCS leakage inside con-
'
tainment (identified during post-trip reviews) resulted in a partial plant cooldown for repairs from December 1-5.
The plant operated briefly on December 6 'until furtheb sec'ondary system equipment failures (moisture separator reheater baffles dislodged) were identified. The plant was shut down for repairs until December 11. After plant startup on December
.12, power was held at 16 percentdecause the licensee identified a prob-lem that could' affect the operability of the ECCS systems in the high-
' head recirculation mode following certain medium-sized-break LOCAs. On December 19, the plant shut down to perform a special flow test on the ECCS system to verify the adequacy of the proposed problem resolution.
Following ' successful completion of the test on December 20, and NRC
. approval of the licensee's'corYective actions, the plant was restarted on December 23 and reached full power on December 25, 1986. With the exception of brief power reductions for routine testing and maintenance, full power operation continuec' through the end of the SALP period (March 31,.1987).
r.
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_ _ _ _. _ _ _ _ _ _ _ - _ _ - - -
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. ih TABLE 1 INSPECTION REPORT ACTIVITIES REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
,
213/86-03 RESIDENT 469 ROUTINE RESIDENT 2/11-4/15/86
<
213/86-04 SPECIALIST
RADWASTE TRANSPORTATION 3/10-14/86 213/86-06 RESIDENT 138 ROUTINE RESIDENT 4/15-5/27/86 213/86-07 SPECIALIST
PDCR MANAGEMENT MEETING 3/26/86 213/86-08 SPECIALIST
CONTAINMENT INTEGRATED LEAK RATE TEST 3/10-14/86 (CILRT) PREPARATIONS
213/86-09 SPECIALIST
CILRT AND RESULTS 3/31-4/16/86 213/86-10 CANCELLED 213/86-11 SPECIALIST 120 ALARA PROGRAM APPRAISAL 4/7-))/86 213/86-12 SPECIALIST 150 1986 EMERGENCY EXERCISE 4/25-26/86 213/86-13 SPECIALIST
INSERVICE TEST PROGRAM REVIEW 4/29-5/2/86 213/86-14 SPECIALIST
OPERATOR REQUALIFICATION 6/16-17/86 213/86-15 SPECIALIST
RADI0 ACTIVE EFFLUENTS 5/19-22/86 213/86-16 RESIDENT 156 ROUTINE RESIDENT 5/28-7/8/86 v
213/86-17 SPECIALIST 170 APPENDIX R IMPLEMENTATION 6/16-20/86 213/86-18 CANCELLED
!
T1-1
_ ______
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REPORT / DATES INSPECTOR HOURS AREAS INSPECTED L
213/86-19 SPECIALIST
MAINTENANCE PROGRAM 7/7/86-11/86 213/86-20 RESIDENT 142 ROUTINE RESIDENT 7/8-8/14/86 213/86-21 CANCELLED 213/86-22 SPECIALIST
FOLLOWUP ON PERSONNEL OVEREXPOSURE 7/22-25/86 213/86-23 SPECIALIST
'STARTUP TESTING 7/28-8/1/86 213/86-24 RESIDENT 137 ROUTINE RESIDENT 8/15-9/30/86 213/86-25 SPECIALIST 100 MASONRY WALL DESIGN 10/21-24/86 213/86-26 SPECIALIST OPERATOR LICENSE EXAMS
11/3-7/86-213/86-27 RESIDENT 261 ROUTINE RESIDENT 10/1-11/17/86 213/86-28 SPECIALIST
NON-LICENSED STAFF TRAINING 11/3-7/86 213/86-29 SPECIALIST 509 EVALUATIVE TEAM INSPECTION 11/14-21/86 213/86-30 RESIDENT 158 ROUTINE RESIDENT 11/18-12/17/86 213/87-01 SPECIALIST
CALIBRATION 1/6-9/87 213/87-02 RESIDENT 229 ROUTINE RESIDENT 12/18/86-2/9/87 213/87-03 SPECIALIST
EMERGENCY PREPAREDNESS 2/2-05/87 213/87-04 SPECIALIST
PHYSICAL SECURITY 1/27-28/87 TI-2
_ _ - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ -
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REPORT / DATES INSPECTOR'
HOURS ARhAS INSPECTED
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'213/87-05-SPECIALIST
' 35 FIRE PROTECTION FOL.LOWUP
-
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RESIDENT 225 30' TINE RESIDENT J
.I 2h,D-3/18/87
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'* This report documents an operator licensing exam # nation. There were no inspection related hours.
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TABLE 2 INSPECTION HOUR SUMMARY l
AREA liOURS HR/YR
% OF TIME OPERATIONS 1270 1172 35.4
,
l MAINTENANCE 426 393 11.9 SURVEILLANCE 460 425 12.8 FIRE PROTECTION 276 255 7.7 l
l-;
LICENSING
--
--
0.0 OUTAGEf 240 222 6.7 RAD PROTECTION 507 468 14.1 I
EMERGENCY PREP.
201 186 5.6 SEC/ SAFEGUARDS 188 173 5.2 TRAINING-0.0
--
--
ASSURANCE OF QUALITY
20 0.6
-
TOTALS:
3590 3314 100.0
_.
T2-1
_
- _ - - - _ _ _ - - _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ - - _ _ _ _ _ - _ _ _ _ _ _ _
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TABLE _3_
' ENFORCEMENT SUMMARY SEVERITY LEVEL AREA
2
4
DEV TOTAL 10PERATIONS
1
MAINTENANCE
1 3-SURVEILLANCE
1
'4 FIRE. PROTECTION-2
1
LICENSING OUTAGES RAD PROTECTION
2
4
= EMERGENCY PREP.
SEC/ SAFEGUARDS TRAINING ASSURANCE OF QUALITY-TOTALS:
12
1
INSPECTION VIOL.
FUNCTIONAL REPORT /DATE REQUIREMENT LEVEL AREA VIOLATION 213/86-03'
TS 6.8-5 MAINTENANCE LICENSEE FAILED TO IMPLEMENT 2/11-4/15/86 PROCEDURES FOR STEAM GENERATOR WET LAYUP MODIFICATION'AT SYSTEM TURNOVER IN ACCORDANCE WITH PROCEDURE 1.2-3.1 213/86-04 10 CFR'50 4*
RAD-CHEM RADWASTE PROCEDURES ACCEPTANCE
~ /10-14/86 APP B, CRS CRITERIA DID NOT PROVIDE FOR
-
VERIFICATION OF RADIONUCLIDES IN SOLID WASTE SHIPMENTS 213/86-04 10 CFR 2.311B 4*
RAD-CHEM RADWASTE SHIPMENT MANIFESTS DID
'3/10-14/86 NOT IDENTIFY IRON 55 AS A CON-TAINED RADIONUCLIDES IN VARIOUS SHIPMENTS 213/86-04 10 CFR 20.311 4*
RAD-CHEM INADEQUATE CERTIFICATION OF RAD-
,
03/10-14/87'
C'
WASTE SHIPMENT DESCRIPTION 213/86-04 10 CFR 71.5 4*
RAD-CHEM SHIPPING PAPERS DID NOT INCLUDE 03/10-14/87 49 CFR 172 IRON 55 CONTAINED IN MANY SHIP-MENTS T3-1 L
-
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--
-
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_ - _ _ _ _ _ _ _ __-
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INSPECTION VIOL.
FUNCTIONAL REPORT /DATE REQUIREMENT LEVEL-AREA VIOLATION-
'213/86-08 TS 4.4
.4 SURVEILLANCE LOCAL LEAK RATE TESTING USING L
03/10-14/87 AN. UNAUTHORIZED TEST METHOD-
'213/86-08
SURVEILLANCE FAILURE TO DOCUMENT A SAFETY 03/10-14/86 EVALUATION FOR CHANGES TO THE CONTAINMENT BOUNDARIES 213/86-08 1TS 3.11
OPERATIONS BREACH OF CONTAINMENT INTEGRITY-03/10-14/86 ON 3 OCCASIONS
- 213/86-15 TS 8.~1.1.1'
RAD-CHEM LICENSEE FAILED TO PERFORM 5/19-22/86 CHEMICAL ANALYSES AT THE PRE-SCRIBED FREQUENCY.
- 213/86-15 TS 6.8
RAD-CHEM CHEMISTRY PROCEDURES DID NOT PRE-
'5/19-22/86 SCRIBE ALL REQUIRED ANALYSES FOR LIQUID RELEASES
- 213/86-17 10 CFR 50
FIRE PROT.
CONTROL ROOM HALON SUPPRESSION 6/16-20/86 APP R SYSTEM TEST DID NOT MEET COM-MITTED ACCEPTANCE CRITERIA.
213/86-17.
10 CFR'50
FIRE PROT.
ALTERNATE SHUTDOWN /C00LDOWN 6/16-20/86 APP R ANALYSIS INADEQUATE BECAUSE OF ERROR IN STEAM GENERATOR-VENT CAPACITY.
213/86-17-TS'6.8.1
FIRE PROT.
INADEQUATE BREAKER COORDINATION 6/16-20/86 SETTING PROCEDURES.
s 213/86-17 10 CFR'50
FIRE PROT.
INADEQUATE EMERGENCY LIGHTING 6/16-20/86 APP R FOR THE B CHARGING PUMP.
213/86-17 CYAPCo LTR DEV FIRE PROT.
BREAKERS FOR CERTAIN MGTOR OPER-
.6/16-20/86 9/16/86 ATED VALVES WERE NOT LOCKED OPEN AS COMMITTED BY THE LICENSEE.
213/86-19.
MAINTENANCE FAILURE TO PROPERLY DOCUMENT AND 7/7-11/86 CONTROL WORK SCOPE AND MATERIAL ISSUE.
o; 213/86-20 TS 3.11
'4 OPERATIONS MANUAL CONTAINMENT ISOLATION 7/8-8/14/86 VALVES (SI-V-863A,B,C,D) WERE OPENED COMPROMISING CONTAINMENT INTEGRITY AS DEFINED IN TS 3.11.
T3-2
- - - _ _ _ - - _
_ ___ _ _ - _ _
.,.
INSPECTION VIOL.
FUNCTIONAL REPORT /DATE REQUIREMENT LEVEL AREA VIOLATION 213/86-20 TS 6.8.1
FIRE PROT.
CONTROL ROOM FIRE EXTINGUISHERS 7/8-8/14/86 WERE NOT INSPECTED MONTHLY AS REQUIRED BY PLANT PROCEDURES.
213/86-22 10 CFR 20.101 3#
RAD-CHEM CUMULATIVE EXPOSURE OF STEAM 7/22-25/86 GENERATOR WORK IN EXCESS OF 3 REM /QTR.
213/86-22 TS 6.11 3#
RAD-CHEM FAILURE TO CONTROL WORKER EX-7/22-25/86 POSTURE IN ACCORDANCE WITH AD-MINISTRATIVE CONTROL PROCEDURES.
213/86-22 10 CFR 20.201 3#
RAD-CHEM FAILURE TO PERFORM ADEQUATE SUR-7/22-25/86 VEYS TO EXTEND WORKER STAY TIME IN A HIGH RADIATION AREA.
213/86-27 TS 3.11
OPERATIONS UNAUTHORIZED OPERATION OF CON-10/1/-11/17/86 TAINMENT ISOLATION VALVE CC-V-884.
213/86-27 10 CFR 50
MAINTENANCE INADEQUATE POST-MAINTENANCE 10/1-11/17/86 APP. J TESTING FOR VALVE CC-CV-885.
213/86-27 TS 6.8
OPERATIONS FAILURE TO FOLLOW SURVEILLANCE 10/1-11/17/86 PROCEDUPE SUR-5.1.4 SUCH THAT VALVE Si-V-865 WAS NOT RECLOSED UPON TEST COMPLETION.
213/86-27 TS 6.8
SURVEILLANCE FAILURE TO FOLLOW SUR 5.7-19 SUCH 10/1-11/17/86 THAT PRESSURE GAUGES WERE NOT CALIBRATED AND DATA SHEETS NOT ATTACHED FOR TEST.
213/87-01 10 CFR 50
SURVEILLANCE INADEQUATE ADMINISTRATIVE CON-1/6-9/87 APP B TROLS OVER SUBSTITUTION / CHANGE OF TEST METHODS AND EQUIPMENT.
- These four violations were cited as an aggregate Severity Level 4 violation.
- Three violations were cited as an aggregate Severity Level 3 violation.
l T3-3 L__-_ _ _
_ _ _ _ _.. _ _ _ _ _
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I TABLE 4
'
LISTING OF LERs BY FUNCTIONAL AREA AREA A
B C
D E
X TOTAL OPERATIONS
4
2
MAINTENANCE
1
SURVEILLANCE
2
9 FIRE PROTECTION
1
7 LICENSING OUTAGES RAD PROTECTION
2 EMERGENCY PREP.
SEC/ SAFEGUARDS TRAINING
/tSSURANCE OF QUALITY TOTALS:
T7
15
4f Cause Codes *:
A - Personnel Error B - Design, Manufacturing, Construction, or Installation Error C - External Cause D - Defective Procedures E - Component Failure X - Other
- Cause Codes in this table are based on inspector evaluations and may differ from those specified in the LER.
T4-1
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _
- _ _
. - _ _.
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- - - _
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_ _ _ _.
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LER.
EVENT CAUSE NUMBER DATE CODE DESCRIPTION 86-10 2/26/86 A
UNAUTHORIZED ENTRANCE TO A HIGH RADIATION AREA.
86-11 2/28/86.
A INOPERABLE FIRE SPRINKLER SYSTEM.
86-12 2/26/86.
X DROPPED FUEL ELEMENT EVENT.
86-13 3/25/86
INADEQUATE SMALL BREAK LOSS OF COOLANT ACCIDENT ANALYSIS.
86-14 3/8/80 l
INOPERABLE FIRE WATER PUMPS 86-15-3/19/86 E
CONTROL ROD WEAR AND CRACKING.
86-16 3/12/86 E
INOPERABLE C02 FIRE PROTECTION SYSTEM.
86-17 3/17/86 A
DEGRADED FIRE BARRIER SEALS.
86-18 4/9/86 B
REACTOR COOLANT SYSTEM WIDE RANGE PRESSURE UNCERTAINTY 86-19 5/6/86 A
DEFECTIVE STEAM GENERATOR TUBE NOT PLUGGED 86-20.
5/8/86 A
REACTOR TRIP DURING TURBINE VIBRATION TESTING 86-21 5/24/86 E
NUCLEAR INSTRUMENTATION DROPPED R0D SETPOINT DRIFT 86-P2 5/17/86 A
TURBINE RUNBACK DURING NUCLEAR INSTRUMENTATION ADJUSTMENT 86-23 5/18/86 E
SPURIOUS LOAD RUNBACK ACTUATIONS 86-24 6/9/86 E
SPURIOUS TURBINE LOAD RUNBACK 86-25 6/19/86 A
REACTOR TRIP DURING NUCLEAR INSTRUMENTATION TROUBLESHOOTING 86-26 6/22/86 X
LOW FLOW TRIP CAUSED BY FOREIGN MATERIAL IN REACTOR PROTECTION SYSTEM 86-27 6/4/86 E
REACTOR TRIPS DUE TO LOSS OF HEATER DRAIN PUMP FLOW 86-28 6/10/86 A
INOPERABLE SWITCHGEAR HALON FIRE PROTECTION I
SYSTEM
l l
l T4-2
.-
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ -
.,
e LER EVENT CAUSE NUMBER DATE CODE DESCRIPTION 86-29 6/17/86 B
MAIN STEAM ISOLATION VALVE STROKE TEST FAILURE 86-30 6/27/86 E
SPURIOUS LOAD RUNBACK 86-31 7/3/86 A
HIGH PRESSURE SAFETY INJECTION RECIRCULATION VALVE FAILURE 86-32 7/10/86 E
DROPPED ROD PROTECTION SETPOINT DRIFT 86-33 7/15/86 B
LOW TEMPERATURE OVERPRESSURIZATION PROTECTION ISOLATION VALVE INTERLOCK FAILURE 86-34 7/11/86 B
INADEQUATE 3-LOOP REACTOR COOLANT SYSTEM FLOW-RATE 86-35 7/15/86 E
MULTIPLE STEAM GENERATOR TUBE DEFECTS (MINI-OUTAGE)
86-36 7/8/86 X
CHARGING PUMP SHAFT FAILURE-86-37 7/8/86 A
MISOPERATION OF CONTAINMENT ISOLATION VALVES 86-38 8/6/86 A
INADEQUATE RETEST OF NUCLEAR INSTRUMENTATION 86-39 7/23/86 A
RADIATION WORKER OVEREXPOSURE 86-40 7/24/86 A
LOW TEMPERATURE OVERPRESSURIZATION PROTECTION RENDERED INOPERABLE FOR TESTING 86-41 8/30/86 E
REACTOR TRIP CAUSED BY FAILED FEED REGULATING VALVE 86-42 10/16/86 A
OPERATOR SURVEILLANCE MISSED INOPERABLE LOCK ON SAFETY INJECTION VALVE 86-43 11/8/86 E
WIDE RANGE STACK MONITOR FAILED LOW 86-44 11/30/86 A
AUTOMATIC REACTOR TRIP DUE TO PERSONNEL ERROR DURING MAINTENANCE 86-45 11/14/86 B
TURBINE SPRINKLER SYSTEM DISABLED TO PERFORM APPENDIX R MODIFICATIONS l
86-46 12/4/86 E
LOW TEMPERATURE OVERPRESSURIZATION PROTECTION l
RELIEF VALVE FAILED TO RESEAT FOLLOWING IN-ADVERTENT RCS PRESSURE SPIKE T4-3 E 1-
.
__
,
_ - - -, - - -, -. - - - -,. - - -, - - - - -
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.9 LER EVENT
'CAUSE NUMBER DATE CODE DESCRIPTION 86-47 12/8/87 A
PLANT IN UNANALYZED CONDITION DUE TO ADMINI-STRATIVE ERROR 86-48 12/12/86 B
EMERGENCY CORE COOLING SYSTEM PUMP RUN0VT DURING CORE DELUGE LINE BREAK 87-01 1/17/87 A
FIRE WATCH DEPARTED WITHOUT RELIEF.
87-02 03/02/87 E
SPURIOUS TURBINE LOAD RUNBACK.
87-03 03/06/87 E
FAILURE OF THE WIDE RANGE STACK MONITOR.
T4-4
_ _ _ _ _ _ _ _ _ _
l
- _ - - _
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.
e TABLE 5 SALP HISTORY Ratings for Period Ending AREA 6/81 8/82 8/83 2/85 2/86 Operations
1
1
Radiological Controls
1
2
Maintenance
1
1
Surveillance
1
2
Fire Protection
1
1
2
Security and Safeguards
1
1
Refueling and Outage
1
1
Management Assurance of Quality
2
Licensing
1
2
Training & Qualifications
Effecitvenss
,
,
- No basis available for a rating.
- Not rated as a separate area.
~~
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